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  • 101.
    Gong, Shengjie
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik.
    Simulation and validation of the dynamics of liquid films evaporating on horizontal heater surfaces2012Inngår i: Applied Thermal Engineering, ISSN 1359-4311, E-ISSN 1873-5606, Vol. 48, s. 486-494Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    In this study a non-linear governing equation based on lubrication theory is employed to model the thinning process of an evaporating liquid film and ultimately predict the critical thickness of the film rupture under impacts of various forces resulting from mass loss, surface tension, gravity, vapor recoil and thermo-capillary. It is found that the thinning process in the experiment is well reproduced by the simulation. The film rupture is caught by the simulation as well, but it underestimates the measured critical thickness at the film rupture. The reason may be that the water wettability of the heater surfaces is not taken into account in the model. Thus, the minimum free energy criterion is used to obtain a correlation which combines the contact angle (reflection of wettability) with the critical thickness from the simulation. The critical thicknesses predicted by the correlation have a good agreement with the experimental data (the maximum deviation is less than 10%).

  • 102. Gong, Shengjie
    et al.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Gu, Hanyang
    An experimental investigation on bubble dynamics and boiling crisis in liquid films2014Inngår i: International Journal of Heat and Mass Transfer, ISSN 0017-9310, E-ISSN 1879-2189, Vol. 79, s. 694-703Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    This paper presents an experimental study of boiling and boiling crisis in a liquid film on a heater surface. The critical heat flux (CHF) values obtained in the present experiment mirror that of pool boiling, irrespective of initial liquid film thickness and liquid supply rate in the liquid film boiling case. This observation reinforces to the "scale separation" concept that high-heat-flux boiling and burnout are governed by micro-hydrodynamics in the liquid film on the heater surface. In addition to the CHF data, evolutions of bubbles and dry spots in the boiling liquid film are captured by means of high-speed high-resolution video camera. The dry spots were observed over surface heat flux ranging from 0.3 MW/m(2) to CHF, typically covering an area less than 10% of the heater surface. Three types of dry spot evolution are observed: (1) under the low heat flux, dry spots are rewetted by receding water dam upon rupture of corresponding bubbles; (2) as the heat flux reaches 1.25 MW/m(2), dry spots rewetting is additionally aided by liquid flow driven by growth of bubbles nucleated in the vicinity; (3) upon approaching the CHF, dry spot(s) cannot be rewetted anymore and expand laterally, leading to boiling crisis (burnout of the heater surface). The richness of observations and characterization of micro-hydrodynamics in the present study further demonstrates that observations and measurements on boiling liquid films provide a paramount window for investigation and understanding of physical mechanisms of boiling and boiling crisis.

  • 103.
    Gong, Shengjie
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Li, Liangxing
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    An experimental study on boiling phenomena in a liquid layerInngår i: International journal of thermal sciences, ISSN 1290-0729, E-ISSN 1778-4166Artikkel i tidsskrift (Annet vitenskapelig)
  • 104.
    Gong, Shengjie
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Li, Liangxing
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    An experimental study on the effect of liquid film thickness on bubble dynamics2013Inngår i: Applied Thermal Engineering, ISSN 1359-4311, E-ISSN 1873-5606, Vol. 51, nr 1-2, s. 459-467Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Experiments were conducted to investigate the boiling phenomenon in various liquid layers on a silicon heater surface with an artificial cavity. Deionized water is employed as working liquid. The emphasis is placed on how the liquid layer thickness affects bubble behaviour and liquid layer integrity for nucleate boiling under the isolated bubble regime. The experimental results show that for boiling in a liquid layer of ∼7.5 mm, the bubble dynamics reproduce the typical pool boiling characteristics with the averaged maximum diameter of 3.2 mm for the isolated bubbles growing on the cavity. As the water layer thickness decreases to the level comparable with the bubble departure diameter, the bubble is found to remain on the heater surface for an extended period, with a dry spot forming under the bubble but rewetted after the bubble rupture occurs. Further reducing the liquid layer thickness, an irreversible dry spot appears, suggesting a minimum rewettable thickness ranging from 1.2 mm to 1.9 mm corresponding to heat flux of 26 kW/m2 to 52 kW/m2. The void measured in the cavity confirms that it is dry inside the artificial cavity at high heat flux.

  • 105.
    Goronovski, Andrei
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Influence of In-vessel Pressure and Corium Melt Properties on Global Vessel Wall Failure of Nordic-type BWRs2013Independent thesis Advanced level (degree of Master (Two Years)), 20 poäng / 30 hpOppgave
    Abstract [en]

    The goal of the present study is to investigate the effect of different scenarios of core degradation in a Nordic-type BWR (boiling water reactor) on the reactor pressure vessel failure mode and timing. Specifically we consider the effects of (i) in-vessel pressure, (ii) melt properties. Control rod guide tube (CRGT) cooling and cooling of the debris from the top are considered as severe accident management (SAM) measures in this study. We also consider the question about minimal amount of debris that can be retained inside the reactor pressure vessel (RPV).

    Analysis is carried out with coupled (i) Phase-change Effective Convectivity (PECM) model implemented in Fluent for prediction of the debris and melt pool heat transfer, and (ii) structural model of the RPV lower head implemented in ANSYS for simulation of thermo-mechanical creep. The coupling is done through transient thermal load predicted by PECM and applied as a boundary condition in ANSYS analysis.

    Results of the analysis suggest that applying only CRGT and top cooling is insufficient for maintaining vessel integrity with 0.4 m deep (~12 tons) corium melt pool. The failure of the vessel by thermally induced creep can be expected starting from 5.3 h after the dryout of the debris bed in the lower plenum. However, earlier failure of the instrumentation guide tubes (IGTs) is possible due to melting of the nozzle welding.

    The internal pressure in the vessel in the range between 3 to 60 bars has no significant influence on the mode and location of the global RPV wall failure. However, depressurization of the vessel can delay RPV wall failure by 46 min for 0.7 m (~ 30 tons) and by 24 min for 1.9 m (~ 200 tons) debris bed. For 0.7 m pool case, changes in vessel pressure from 3 to 60 bars caused changes in liquid melt mass and superheat from ~18 tons at 180 K to ~13 tons at 100 K superheat, respectively. The same changes in pressure for 1.9 m case caused changes in liquid melt mass and superheat from ~40 tons at 42 K to ~10 tons at about 8 K superheat, respectively.

    Investigation of the influence of melt pool properties on the mode and timing of the vessel failure suggest that the thermo-mechanical creep behavior is most sensitive to the thermal conductivity of solid debris. Both vessel wall and IGT failure timing is strongly dependent on this parameter. For given thermal conductivity of solid debris, an increase in Tsolidus or Tliquidus generally leads to a decrease in liquid melt mass and superheat at the moment of vessel wall failure. Applying models for effective thermal conductivity of porous debris helps to further reduce uncertainty in assessment of the vessel failure and melt ejection mode and timing.

    Only in an extreme case with Tsolidus, Tliquidus range larger than 600 K, with thermal conductivity of solid 0.5 W∙m‑1∙K‑1 and thermal conductivity of liquid melt 20 W∙m‑1∙K‑1, a noticeable vessel wall ablation and melting of the crust on the wall surface was observed. However, the failure was still caused by creep strain and the location of the failure remained similar to other considered cases.

  • 106.
    Goronovski, Andrei
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Tran, Chi-Thanh
    Effect of Corium Non-homogeneity on Nordic BWR Vessel Failure Mode and Timing2015Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Corium melt fragmentation and cooling in a deep pool of water under reactor pressure vessel are employed as severe accident mitigation strategy in a Nordic-type BWR. Core debris relocated to the lower head inflict significant thermal and mechanical loads on the vessel structures. The mode and timing of the vessel failure, mass and superheat of the ejected melt determine ex-vessel accident progression and risks of steam explosion and formation of a non-coolable debris bed. In this work we consider the effect of in-vessel debris non-homogeneity on the mode of vessel failure. The heat-up, re-melting, melt pool formation, and heat transfer of the debris bed are predicted with the Phase-change Effective Convectivity Model (PECM) implemented in FLUENT® code. Then the obtained thermal load on the vessel wall and structures is used as boundary conditions for a thermo-structural analysis of the BWR lower head using the ANSYS® code. In this paper, a corium debris bed is considered inside vessel lower head inducing thermal load on the wall and structures. The debris bed thermal properties axial distribution is taken as a function of material composition, which is extracted from MELCOR® simulations of core failure and debris bed formation inside the lower plenum. A flat and a concave configuration of the debris bed are considered and results of simulations are compared with those for a homogenous debris bed of the same mass-averaged thermal properties.

  • 107.
    Goronovski, Andrei
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    The Effect of Internal Pressure and Debris Bed Thermal Properties on BWR Vessel Lower Head Failure and Timing2013Inngår i: Proc. 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15), 2013Konferansepaper (Fagfellevurdert)
  • 108. Gottlieb, C.
    et al.
    Arzhanov, Vasily
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Gudowski, Waclaw
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Garis, N.
    Feasibility study on transient identification in nuclear power plants using support vector machines2006Inngår i: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 155, nr 1, s. 67-77Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Support vector machines (SVMs), a relatively new paradigm in statistical learning theory, are studied for their potential to recognize transient behavior of detector signals corresponding to various accident events at nuclear power plants (NPPs). Transient classification is a major task for any computer-aided system for recognition of various malfunctions. The ability to identify the state of operation or events occurring at an NPP is crucial so that personnel can select adequate response actions. The Modular Accident Analysis Program, version 4 (MAAP4) is a program that can be used to model various normal and abnormal events in an NPP. This study uses MAAP signals describing various loss-of-coolant accidents in boiling water reactors. The simulated sensor readings corresponding to these events have been used to train and test SVM classifiers. SVM calculations have demonstrated that they can produce classifiers with good generalization ability for our data. This in, turn indicates that SVMs show promise as classifiers for the learning problem of identifying transients.

  • 109. Granovsky, V. S.
    et al.
    Khabensky, V. B.
    Krushinova, E. V.
    Vitol, S. A.
    Sulatsky, A. A.
    Almjashev, V. I.
    Bechta, Sevostian V.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Gusarov, V. V.
    Barrachin, M.
    Bottomle, P. D.
    Fischer, M.
    Piluso, P.
    Oxidation effect on steel corrosion and thermal loads during corium melt in-vessel retention2014Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 278, s. 310-316Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    During a severe accident with core meltdown, the in-vessel molten core retention is challenged by the vessel steel ablation due to thermal and physicochemical interaction of melt with steel. In accidents with oxidizing atmosphere above the melt surface, a low melting point UO2+x-ZrO2-FeOy corium pool can form. In this case ablation of the RPV steel interacting with the molten corium is a corrosion process. Experiments carried out within the International Scientific and Technology Center's (ISTC) METCOR Project have shown that the corrosion rate can vary and depends on both surface temperature of the RPV steel and oxygen potential of the melt. If the oxygen potential is low, the corrosion rate is controlled by the solid phase diffusion of Fe ions in the corrosion layer. At high oxygen potential and steel surface layer temperature of 1050 degrees C and higher, the corrosion rate intensifies because of corrosion layer liquefaction and liquid phase diffusion of Fe ions. The paper analyzes conditions under which corrosion intensification occurs and can impact on in-vessel melt retention (IVR).

  • 110. Granovsky, V. S.
    et al.
    Sulatsky, A. A.
    Khabensky, V. B.
    Sulatskaya, M. B.
    Gusarov, V. V.
    Almyashev, V. I.
    Komlev, A. A.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kim, Y. S.
    Park, R. J.
    Kim, H. Y.
    Song, J. H.
    Modeling of melt retention in EU-APR1400 ex-vessel core catcher2012Inngår i: International Congress on Advances in Nuclear Power Plants 2012, ICAPP 2012, 2012, s. 1412-1421Konferansepaper (Fagfellevurdert)
    Abstract [en]

    A core catcher is adopted in the EU-APR1400 reactor design for management and mitigation of severe accidents with reactor core melting. The core catcher concept incorporates a number of engineering solutions used in the catcher designs of European EPR and Russian WER-1000 reactors, such as thin-layer corium spreading for better cooling, retention of the melt in a water-cooled steel vessel, and use of sacrificial material (SM) to control the melt properties. SM is one of the key elements of the catcher design and its performance is critical for melt retention efficiency. This SM consists of oxide components, but the core catcher also includes sacrificial steel which reacts with the metal melt of the molten corium to reduce its temperature. The paper describes the required properties of SM. The melt retention capability of the core catcher can be confirmed by modeling the heat fluxes to the catcher vessel to show that it will not fail. The fulfillment of this requirement is demonstrated on the example of LBLOCA severe accident. Thermal and physicochemical interactions between the oxide and metal melts, interactions of the melts with SM, sacrificial steel and vessel, core catcher external cooling by water and release of non-condensable gases are modeled.

  • 111.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Basso, Simone
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Galushin, Sergey
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Development of Texas-V code surrogate model for assessment of steam explosion impact in Nordic BWR2015Inngår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, American Nuclear Society, 2015, Vol. 9, s. 7222-7235Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Severe accident mitigation strategies in Nordic boiling water reactors (BWRs) employ core melt cooling in a deep pool of water under the reactor pressure vessel. Corium melt released from the vessel is expected to fragment, solidify and form a porous debris bed coolable by natural circulation. However, steam explosion can occur upon melt release threatening containment integrity and potentially leading to large early release of radioactive products to the environment. Significant aleatory and epistemic uncertainties exist in accident scenarios, melt release conditions, and modeling of steam explosion phenomena. Assessment of the risk of ex-vessel steam explosion requires application of the Integrated Deterministic Probabilistic Safety Analysis (IDPSA). IDPSA is a computationally demanding task which makes unfeasible direct application of Fuel-Coolant Interaction codes. The goal of the current work is to develop a Surrogate Model (SM) of the Texas-V code and demonstrate its application to the analysis of explosion impact in the Nordic BWR. The SM should be computationally affordable for IDPSA analysis. We focus on prediction of the steam explosion loads in a reference Nordic BWR design assuming a scenario of coherent corium jet release into a deep water pool. We start with the review of the Texas-V sub-models in order to identify a list of parameters to be considered in implementation of the SM. We demonstrate that Texas-V exhibits chaotic response in terms of the explosion impulse as a function of the triggering time and introduce a statistical representation of the explosion impulse for given melt release conditions and arbitrary triggering time. We demonstrate that characteristics of the distribution are well-posed. We then separate out the essential portion of modelling uncertainty by identification of the most influential uncertain parameters using sensitivity analysis. Both aleatory uncertainty in characteristics of melt release scenarios and water pool conditions, and epistemic uncertainty in FCI modeling are considered. Ranges of the uncertain parameters are selected based on the available information about prototypic severe accident conditions in a Nordic BWR. A database of Texas-V solutions is generated and used for the development of the SM. Performance, predictive capability and application of the SM to risk analysis are discussed in detail.

  • 112.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Basso, Simone
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Development of a surrogate model for analysis of ex-vessel steam explosion in Nordic type BWRs2016Inngår i: NUCLEAR ENGINEERING AND DESIGN, ISSN 0029-5493, Vol. 310, s. 311-327Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Severe accident mitigation strategy adopted in Nordic type Boiling Water Reactors (BWRs) employs ex vessel core melt cooling in a deep pool of water below reactor vessel. Energetic fuel coolant interaction (steam explosion) can occur during molten core release into water. Dynamic loads can threaten containment integrity increasing the risk of fission products release to the environment. Comprehensive uncertainty analysis is necessary in order to assess the risks. Computational costs of the existing fuel coolant interaction (FCI) codes is often prohibitive for addressing the uncertainties, including the effect of stochastic triggering time. This paper discusses development of a computationally efficient surrogate model (SM) for prediction of statistical characteristics of steam explosion impulses in Nordic BWRs. The TEXAS-V code was used as the Full Model (FM) for the calculation of explosion impulses. The surrogate model was developed using artificial neural networks' (ANNs) and the database of FM solutions. Statistical analysis was employed in order to treat chaotic response of steam explosion impulse to variations in the triggering time. Details of the FM and SM implementation and their verification are discussed in the paper.

  • 113.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Basso, Simone
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Sensitivity Study of Steam Explosion Characteristics to Uncertain Input Parameters Using TEXAS-V Code2014Inngår i: NUTHOS10, Paper-1293, Okinawa, Japan, 2014, Atomic Energy Society of Japan , 2014Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Release of core melt from failed reactor vessel into a pool of water is adopted in several existing designs of light water reactors (LWRs) as an element of severe accident mitigation strategy. Corium melt is expected to fragment, solidify and form a debris bed coolable by natural circulation. However, steam explosion can occur upon melt release threatening containment integrity and potentially leading to large early release of radioactive products to the environment. There are many factors and parameters that could be considered for prediction of the fuel-coolant interaction (FCI) energetics, but it is not clear which of them are the most influential and should be addressed in risk analysis. The goal of this work is to assess importance of different uncertain input parameters used in FCI code TEXAS-V for prediction of the steam explosion energetics. Both aleatory uncertainty in characteristics of melt release scenarios and water pool conditions, and epistemic uncertainty in modeling are considered. Ranges of the uncertain parameters are selected based on the available information about prototypic severe accident conditions in a reference design of a Nordic BWR. Sensitivity analysis with Morris method is implemented using coupled TEXAS-V and DAKOTA codes. In total 12 input parameters were studied and 2 melt release scenarios were considered. Each scenario is based on 60,000 of TEXAS-V runs. Sensitivity study identified the most influential input parameters, and those which have no statistically significant effect on the explosion energetics. Details of approach to robust usage of TEXAS-V input, statistical enveloping of TEXAS-V output and interpretation of the results are discussed in the paper. We also provide probability density function (PDF) of steam explosion impulse estimated using TEXAS-V for reference Nordic BWR. It can be used for assessment of the uncertainty ranges of steam explosion loads for given ranges of input parameters.

  • 114.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Galushin, Sergey
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Basso, Simone
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Failure domain analysis and uncertainty quantification using surrogate models for steam explosion in a nordic type BWR2017Inngår i: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery, Inc , 2017Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Sever accident mitigation strategy adopted in Nordic Boiling Water Reactors (BWRs) employs a deep water pool below the reactor vessel in order to fragment and quench core melt and provide long term cooling of the debris. One of the risk factors associated with this accident management strategy is early failure of the containment due to steam explosion. Assessment of the risk is subject to significant epistemic and aleatory uncertainties in (i) modelling of steam explosion and (ii) scenarios of melt release from the vessel and water pool conditions. High computational efficiency of the models is required for such assessment. A surrogate model (SM) approach has been previously developed using artificial neural network and the database of Texas-V code solutions for steam explosion loads in the Nordic type BWRs. In this paper we extend our surrogate model to allow analysis of steam explosion in relatively shallow water pools (>2 m), address effects of melt emissivity and resolve more accurately variation of pressure in the drywell. We provide detailed comparison of metallic vs oxidic melt release scenarios, incorporate uncertainty of the SM into modelling and analyze the sensitivity of our results to SM uncertainty. We estimate risks of containment failure with non-reinforced and reinforced hatch door and demonstrate the effect of the surrogate model uncertainty on the results. We analyze the results and develop a simplified approach for decision making considering predicted failure probabilities, expected costs and scenario frequencies. 

  • 115.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Jeltsov, Marti
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kööp, Kaspar
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    The TALL-3D facility design and commissioning tests for validation of coupled STH and CFD codes2015Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 290, s. 144-153Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Application of coupled CFD (Computational Fluid Dynamics) and STH (System Thermal Hydraulics) codes is a prerequisite for computationally affordable and sufficiently accurate prediction of thermal-hydraulics of complex systems. Coupled STH and CFD codes require validation for understanding and quantification of the sources of uncertainties in the code prediction. TALL-3D is a liquid Lead Bismuth Eutectic (LBE) loop developed according to the requirements for the experimental data for validation of coupled STH and CFD codes. The goals of the facility design are to provide (i) mutual feedback between natural circulation in the loop and complex 3D mixing and stratification phenomena in the pool-type test section, (ii) a possibility to validate standalone STH and CFD codes for each subsection of the facility, and (iii) sufficient number of experimental data to separate the process of input model calibration and code validation. Description of the facility design and its main components, approach to estimation of experimental uncertainty and calibration of model input parameters that are not directly measured in the experiment are discussed in the paper. First experimental data from the forced to natural circulation transient is also provided in the paper.

  • 116.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Jeltsov, Marti
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kööp, Kaspar
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Design and Commissioning Tests of the TALL-3D Experimental Facility for Validation of Coupled STH and CFD Codes2014Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Application of coupled CFD (Computational Fluid Dynamics) and STH (System Thermal Hydraulics) codes is a prerequisite for computationally affordable and sufficiently accurate prediction of thermal-hydraulics of complex systems. Coupled STH and CFD codes require validation for understanding and quantification of the sources of uncertainties in the code prediction. TALL-3D is a liquid Lead Bismuth Eutectic (LBE) loop developed according to the requirements for the experimental data for validation of coupled STH and CFD codes. The goals of the facility design are to provide (i) mutual feedback between natural circulation in the loop and complex 3D mixing and stratification phenomena in the pool-type test section, (ii) a possibility to validate standalone STH and CFD codes for each subsection of the facility, (iii) sufficient number of experimental data to separate the process of input model calibration and code validation. Description of the facility design and its main components, approach to estimation of experimental uncertainty and calibration of model input parameters that are not directly measured in the experiment are discussed in the paper. First experimental data from the forced to natural circulation transient is also provided in the paper.

  • 117.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Konovalenko, Alexander
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinova, Valtyna
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Insight into steam explosion in stratified melt-coolant configuration2013Inngår i: 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, 2013Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Release of core melt from failed reactor vessel into a pool of water is adopted in several existing designs of light water reactors (LWRs) as an element of severe accident  mitigation  strategy.  When  vessel  breach  is  large  and  water  pool  is shallow,  released  corium  melt  can  reach  containment  floor  in  liquid  form  and spread under water creating a stratified configuration of melt covered by coolant. Steam  explosion  in  such  stratified  configuration  was  long  believed  as  of secondary importance for reactor safety because it was assumed that considerable mass of melt cannot be premixed with the coolant. In this work we revisit these assumptions  using  recent  experimental  observations  from  the  stratified  steam explosion tests  in  PULiMS  facility.  We  demonstrate  that  (i)  considerable  melt-coolant premixing layer can be formed in the stratified configuration with high temperature  melts,  (ii)  mechanism  responsible  for  the  premixing  is  apparently more  efficient  than  previously  assumed  Rayleigh-Taylor  or  Kelvin-Helmholtz instabilities.  We  also  provide  data  on  measured  and  estimated  impulses, energetics  of  steam  explosion,  and  resulting  thermal  to  mechanical  energy conversion ratios. 

  • 118.
    Grishchenko, Dmitry
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Mickus, Ignas
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    Experimental activities report on TALL-3D2015Rapport (Annet vitenskapelig)
  • 119. Gu, H.
    et al.
    Wang, C.
    Gong, S.
    Mei, Y.
    Li, H.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Investigation on contact angle measurement methods and wettability transition of porous surfaces2016Inngår i: Surface & Coatings Technology, ISSN 0257-8972, E-ISSN 1879-3347, Vol. 292, s. 72-77Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Various solid surfaces (e.g., smooth titanium surface, smooth aluminum surface, polished copper surfaces, polished silver surfaces and porous copper surfaces) were prepared to quantify the reliability of half-angle algorithm and axisymmetric drop shape analysis (ADSA) algorithm for calculating contact angles. Besides, the effects of surface conditions on contact angle values were also investigated. The experimental results of 10 repeated tests for each surface show that both algorithms have good accuracy for an acute contact angle, while the ADSA algorithm is better than the half-angle algorithm for an obtuse contact angle. Furthermore, with the decrease of surface roughness, the contact angle increases but the standard deviation of contact angles by 10 repeated tests decreases. In addition, the porous layer on copper surface by electrochemical deposition shows a super hydrophilic property, but it could change to be super hydrophobic after exposed in ambient air for 24 h. Interestingly, the surface wettability reverses to be super hydrophilic again after it is immersed in water, and the inorganic contamination is the reason of formal change from the super hydrophilic status to the super hydrophobic status.

  • 120.
    Gudowski, Waclaw
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Nuclear waste management. Status, prospects and hopes2005Inngår i: Nuclear Physics A, ISSN 0375-9474, E-ISSN 1873-1554, Vol. 752, nr 1-4, s. 623C-632CArtikkel i tidsskrift (Fagfellevurdert)
  • 121.
    Gudowski, Waclaw
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Shvetsov, V.
    Polanski, A.
    Broeders, C.
    The Subcritical Assembly in Dubna (SAD) - Part II: Research program for ADS-demo experiment2006Inngår i: Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, ISSN 0168-9002, E-ISSN 1872-9576, Vol. 562, nr 2, s. 887-891Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Subcritical Assembly in Dubna (SAD), a project funded by the International Science and Technology Centre, driven in collaboration with many European partners, may become the first Accelerator Driven Subcritical experiment coupling an existing proton accelerator of 660 MeV with a compact MQX-fuelled subcritical core. The main objective of the SAD experiment is to study physics of Accelerator Driven System ranging from a very deep subcriticality up to k(eff) of 0.98. All experiences with subcriticality monitoring from previous subcritical experiments like MUSE, Yalina and IBR-30 booster mode will be verified in order to select the most reliable subcriticality monitoring technique. Particular attention will be given to validation of the core power-beam current relation. Moreover, some studies have been done to assess possibility of power upgrade for SAD.

  • 122.
    Hansson Concilio, Roberta
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    TRIGGERING AND ENERGETICS OF A SINGLE DROPVAPOR EXPLOSION: THE ROLE OF ENTRAPPED NONCONDENSABLE GASES2009Inngår i: NUCLEAR ENGINEERING AND TECHNOLOGY, ISSN 1738-5733, Vol. 41, nr 9, s. 1215-1222Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The present work pertains to a research program to study Molten Fuel-Coolant Interactions (MFCI), which may occur in anuclear power plant during a hypothetical severe accident. Dynamics of the hot liquid (melt) droplet and the volatile liquid (coolant)were investigated in the MISTEE (Micro-Interactions in Steam Explosion Experiments) facility by performing well-controlled,externally triggered, single-droplet experiments, using a high-speed visualization system with synchronized digital cinematographyand continuous X-ray radiography.The current study is concerned with the MISTEE-NCG test campaign, in which a considerable amount of non-condensablegases (NCG) are present in the film that enfolds the molten droplet. The SHARP images for the MISTEE-NCG tests were analyzedand special attention was given to the morphology (aspect ratio) and dynamics of the air/ vapor bubble, as well as the melt droppreconditioning. Energetics of the vapor explosion (conversion ratio) were also evaluated.The MISTEE–NCG tests showed two main aspects when compared to the MISTEE test series (without entrapped air). First,analysis showed that the melt preconditioning still strongly depends on the coolant subcooling. Second, in respect to the energetics,the tests consistently showed a reduced conversion ratio compared to that of the MISTEE test series.

  • 123.
    Hansson Concilio, Roberta
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Effect of non-condensable gases on triggering and energetics of a single drop vapor explosion2008Inngår i: Transactions of the American Nuclear Society, 2008, s. 547-548Konferansepaper (Fagfellevurdert)
  • 124.
    Hansson Concilio, Roberta
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Manickam, Louis
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    The Effect of Binary Oxide Materials on a Single Droplet Vapor Explosion Triggering2011Inngår i: Proceeding of the 14th International Topical Meeting on Nuclear Reactor Thermalhydraulics (NURETH-14, 2011Konferansepaper (Fagfellevurdert)
  • 125.
    Hansson Concilio, Roberta
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Park, Hyun sun
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Dynamics and preconditioning in a single drop vapor explosion2007Inngår i: Proceedings - 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-12, 2007Konferansepaper (Fagfellevurdert)
    Abstract [en]

    In order to develop a mechanistic understanding of the thermal-hydraulic processes in vapor explosion, it is paramount to characterize the dynamics of the hot liquid (melt) drop fragmentation and the volatile liquid (coolant) vaporization. In the present study, these intricate phenomena are investigated by performing well-controlled, externally triggered, single-drop experiments, employing a high-speed digital visualization system with synchronized cinematography and X-ray radiography system called SHARP (Simultaneous High-speed Acquisition of X-ray Radiography and Photography). The processed images, after an elaborate image processing, revealed the internal structure and dynamic evolution of the hot liquid fragmentation and related vaporization of the coolant. Such data gives way to new insights into the physics of the vapor explosion phenomena and quantification of the associated dynamic micro interactions. Analysis of the experimental results shows that, followed an external perturbation (trigger), a high temperature molten material (tin) drop underwent deformation and partial fragmentation already during the first cycle of bubble growth. Analysis of the SHARP data reveals correlation between the drop's dynamics in the first bubble cycle and energetics of the subsequent explosive evaporation in the second cycle. This finding provides a basis to suggest a so-called melt drop preconditioning i.e. deformation/ pre-fragmentation of a hot melt drop immediately following the pressure trigger, being instrumental to the subsequent coolant entrainment and resulting energetics of the so-triggered drop explosion.

  • 126.
    Hansson Concilio, Roberta
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Park, Hyun Sun
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Pre-Conditioning and Dynamic Progression of a Single Drop Vapor Explosion2007Inngår i: proceeding of 12th International Meeting on Nuclear Reactor Thermal Hydraulics 2007 (NURETH 12), Curran Associates, Inc., 2007Konferansepaper (Fagfellevurdert)
  • 127.
    Hansson Concilio, Roberta
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Park, Hyun Sun
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Simultaneous High Speed Digital Cinematographic and X-ray Radiographic Imaging of a Multi-Fluid Interaction with Rapid Phase Changes2007Inngår i: Proceeding of International Conference on Multiphase Flow (ICMF-07), 2007Konferansepaper (Fagfellevurdert)
  • 128.
    Hansson, Roberta Concilio
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Triggering and energetics of a single drop vapor explosion: The role of entrapped non-condensable gases2009Inngår i: Nuclear Engineering and Technology, 2009, Vol. 41, nr 9, s. 1215-1222Konferansepaper (Fagfellevurdert)
    Abstract [en]

    The present work pertains to a research program to study Molten Fuel-Coolant Interactions (MFCI), which may occur in a nuclear power plant during a hypothetical severe accident. Dynamics of the hot liquid (melt) droplet and the volatile liquid (coolant) were investigated in the MISTEE (Micro-Interactions in Steam Explosion Experiments) facility by performing well-controlled, externally triggered, single-droplet experiments, using a high-speed Visualization system with synchronized digital cinematography and continuous X-ray radiography. The current study is concerned with the MISTEE-NCG test campaign, in which a considerable amount of non-condensable gases (NCG) are present in the film that enfolds the molten droplet. The SHARP images for the MISTEE-NCG tests were analyzed and special attention was given to the morphology (aspect ratio) and dynamics of the air/ vapor bubble, as well as the melt drop preconditioning. Energetics of the vapor explosion (conversion ratio) were also evaluated. The MISTEE-NCG tests showed two main aspects when compared to the MISTEE test series (without entrapped air). First, analysis showed that the melt preconditioning still strongly depends on the coolant subcooling. Second, in respect to the energetics, the tests consistently showed a reduced conversion ratio compared to that of the MISTEE test series.

  • 129.
    Hansson, Roberta Concilio
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Park, Hyun Sun
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Dynamics and preconditioning in a single-droplet vapor explosion2009Inngår i: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 167, nr 1, s. 223-234Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The present study aims to develop a mechanistic understanding of the thermal-hydraulic processes in a vapor explosion, which may occur in nuclear power plants during a hypothetical severe accident, involving interactions of high-temperature corium melt and volatile coolant. Dynamics of the hot liquid (melt) droplet and the volatile liquid (coolant) were investigated in the Micro-Interactions in Steam Explosion Experiments (MISTEE) facility by performing well-controlled, externally triggered, single-droplet experiments, using a high-speed visualization system with synchronized digital cinematography and continuous X-ray radiography, called Simultaneous High-speed Acquisition of X-ray Radiography and Photography (SHARP). After an elaborate image processing, the SHARP images depict the evolution of both melt material (dispersal) and coolant (bubble dynamics) and their microscale interactions. The analysis of the data shows a deficiency in using the bubble dynamics alone to provide a consistent explanation of the energetic behavior. In contrast, the SHARP data reveal a correlation between the droplet's dynamics in the bubble's first cycle and the energetics of the subsequent explosive evaporation in the bubble's second cycle. The finding provides a basis to suggest that a so-called melt-droplet preconditioning, i.e., deformation/prefragmentation of a hot melt droplet immediately following the pressure trigger, is instrumental to the subsequent coolant entrainment, evaporation, and energetics of the resulting vapor explosion.

  • 130.
    Hansson, Roberta Concilio
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Park, Hyun Sun
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Simultaneous high speed digital cinematographic and X-ray radiographic imaging of a intense multi-fluid interaction with rapid phase changes2009Inngår i: Experimental Thermal and Fluid Science, ISSN 0894-1777, E-ISSN 1879-2286, Vol. 33, nr 4, s. 754-763Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    As typical for the study of the vapor explosions, the qualitative and quantitative understanding of the phenomena requires visualization of both material and interface dynamics. A new approach to multifluid multiphase visualization is presented with the focus on the development of a synchronized highspeed visualization by digital cinematography and X-ray radiography. The developed system, named SHARP (simultaneous high-speed acquisition of X-ray radiography and photography), and its image processing methodology, directed to an image synchronization procedure and a separate quantification of vapor and molten material dynamics, is presented in this paper. Furthermore, we exploit an intrinsic property of the X-ray radiation, namely the differences in linear mass attenuation coefficients over the beam path through a multi-component system, to characterize the evolution of molten material distribution. Analysis of the data obtained by the SHARP system and image processing procedure developed granted new insights into the physics of the vapor explosion phenomena, as well as, quantitative information of the associated dynamic micro-interactions.

  • 131.
    Henriksson, Krister O. E.
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Nordlund, K.
    Keinonen, J.
    Annihilation of craters: Molecular dynamic simulations on a silver surface2007Inngår i: Physical Review B. Condensed Matter and Materials Physics, ISSN 1098-0121, E-ISSN 1550-235X, Vol. 76, nr 24, s. 15 p.-Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The ability of silver cluster ions containing 13 atoms to fill in a preexisting crater with a radius of about 28 angstrom on a silver (001) target has been investigated using molecular dynamics simulations and the molecular-dynamics-Monte Carlo corrected effective medium potential. The largest lateral distance r between crater and ion was about three times the radius of the preexisting crater, namely, 75 angstrom. The results reveal that when r < 20 angstrom and r>60 angstrom the preexisting crater is partially filled in, and for other distances there is a net growth of the crater. The lattice damage created by the cluster ions, the total sputtering yield, the cluster sputtering yield, and simulated transmission electron microscopy images of the irradiated targets are also presented.

  • 132.
    Henriksson, Krister O. E.
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Nordlund, K.
    Keinonen, J.
    Crater annihilation on silver by cluster ion impacts2007Inngår i: Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, ISSN 0168-583X, E-ISSN 1872-9584, Vol. 255, nr 1, s. 259-264Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Using the MD/MC-CEM potential we have investigated the impacts of 20 keV Ag-13 cluster ions on (001) silver surfaces having one initial crater. This one was made in the zeroth ion impact. The degree of annihilation of the initial crater was investigated as a function of the lateral distance r(i) between the crater and the ion. The impact points were selected randomly inside a circular area with a radius of 75 angstrom centered on the crater. To reduce the total number of simulations, the circular area was divided into annuli. The initial and final atomic positions in the impact simulations were analyzed and the degree of annihilation of the initial crater was determined. The results indicate that for r <= 60 angstrom there is a net growth of the initial crater, and for distances r is an element of (60,80) angstrom there is a small net filling of the crater.

  • 133.
    Henriksson, Krister O. E.
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Sandberg, Nils
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorfysik.
    Wallenius, Janne
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorfysik.
    Carbides in stainless steels: Results from ab initio investigations2008Inngår i: Applied Physics Letters, ISSN 0003-6951, E-ISSN 1077-3118, Vol. 93, nr 19Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The useful properties of steels are due to a complicated microstructure containing iron and chromium carbides. Only some basic physical properties of these carbides are known with high precision, although the carbides may have a vital impact on the performance and longevity of the steel. To improve on this situation, we have performed extensive density-functional theory calculations of several carbides. The quantitative results are in perfect agreement with the relative empirical stability of the carbides. Also, in contradiction with experimental data, we find that Cr23C6 responsible for the hardness of stainless steels is not the most stable chromium-dominated carbide.

  • 134. Hu, X.
    et al.
    Xing, M.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Trace validation against loca transients performed on fix-II facility2013Inngår i: 2013 21st International Conference on Nuclear Engineering: Volume 3: Nuclear Safety and Security; Codes, Standards, Licensing and Regulatory Issues; Computational Fluid Dynamics and Coupled CodesChengdu, China, July 29–August 2, 2013, ASME Press, 2013Konferansepaper (Fagfellevurdert)
    Abstract [en]

    As a latest developed computational code, TRACE is expected to be useful and effective for analyzing the thermal-hydraulic behaviors in design, licensing and safety analysis of nuclear power plant. However, its validity and correctness have to be verified and qualified before its application into industry. Lossof- coolant accident (LOCA) is a kind of transient thermal hydraulic event which has been emphasized a lot as a most important threat to the safety of the nuclear power plant. The FIX-II experiments were performed to produce experimental data for understanding the initial stage of LOCA and so as to verify the computational codes. In the present study, based on FIX- II LOCA tests, simulation models for the tests of No. 3025, No. 3061 and No. 5052 which correspond to different LOCA cases were developed to validate the TRACE code (version 5.0 patch 2). The predictions of the TRACE code including the pressure in the primary system, the mass flow rate in certain key parts, and the temperature in the core were compared with the experimental data. The results show that TRACE model can well reproduce the transient thermalhydraulic behaviors under different LOCA situations. In addition, sensitivity analysis are also performed to investigate the influence of particular models and parameters, including counter current flow limitation (CCFL) model and choked flow model on the results, which show that both the models have significant influence on the outcome of the model.

  • 135.
    HU, XIAO
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    TRACE Analysis of LOCA Transients Performed on FIX-II Facility2012Independent thesis Advanced level (degree of Master (Two Years)), 20 poäng / 30 hpOppgave
    Abstract [en]

    As a latest developed computational code, TRACE is expected to be useful and effective for analyzing the thermal-hydraulic behaviors in design, licensing and safety analysis of nuclear power plant. However, its validity and correctness have to be verified and qualified before its application into industry. Loss-of-coolant accident (LOCA) is a kind of transient thermal hydraulic event which has been emphasized a lot as a most important threat to the safety of the nuclear power plant. In the present study, based on FIX- II LOCA tests, simulation models for the tests of No. 3025, No. 3061 and No. 5052 were developed to validate the TRACE code (version 5.0 patch 2). The simulated transient thermal-hydraulic behaviors during the LOCA tests including the pressure in the primary system, the mass flow rate in certain key parts, and the temperature in the core are compared with experimental data. The simulation results show that TRACE model can well reproduce the transient thermal-hydraulic behaviors under different LOCA situations. In addition, sensitivity analysis are also performed to investigate the influence of particular models and parameters, including counter current flow limitation (CCFL) model, choked flow model , insulator in the steam dome, K-factor in the test section, and pump trip, on the results. The sensitivity analyses show that both the models and parameters have significant influence on the outcome of the model.

  • 136.
    Hua, Li
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Validation of Effective Momentum and Heat Flux Models for Stratification and Mixing in a Water Pool2013Rapport (Annet vitenskapelig)
    Abstract [en]

    The pressure suppression pool is the most important feature of the pressure suppression system in a Boiling Water Reactor (BWR) that acts primarily as a passive heat sink during a loss of coolant accident (LOCA) or when the reactor is isolated from the main heat sink. The steam injection into the pool through the blowdown pipes can lead to short term dynamic phenomena and long term thermal transient in the pool. The development of thermal stratification or mixing in the pool is a transient phenomenon that can influence the pool's pressure suppression capacity. Different condensation regimes depending on the pool's bulk temperature and steam flow rates determine the onset of thermal stratification or erosion of stratified layers. Previously, we have proposed to model the effect of steam injection on the mixing and stratification with the Effective Heat Source (EHS) and the Effective Momentum Source (EMS) models. The EHS model is used to provide thermal effect of steam injection on the pool, preserving heat and mass balance. The EMS model is used to simulate momentum induced by steam injection in different flow regimes. The EMS model is based on the combination of (i) synthetic jet theory, which predicts effective momentum if amplitude and frequency of flow oscillations in the pipe are given, and (ii) model proposed by Aya and Nariai for prediction of the amplitude and frequency of oscillations at a given pool temperature and steam mass flux. The complete EHS/EMS models only require the steam mass flux, initial pool bulk temperature, and design-specific parameters, to predict thermal stratification and mixing in a pressure suppression pool. In this work we use EHS/EMS models implemented in containment thermal hydraulic code GOTHIC. The PPOOLEX experiments (Lappeenranta University of Technology, Finland) are utilized to (a) quantify errors due to GOTHIC's physical models and numerical schemes, (b) propose necessary improvements in GOTHIC sub-grid scale modeling, and (c) validate our proposed models. The data from PPOOLEX STR-06, STR-09 and STR-10 tests are used for validation of the EHS and EMS models in this work. We found that estimations of the amplitude and frequency based on available experimental data from PPOOLEX experiments STR-06, STR-09, and STR-10 have too large uncertainties due to poor space and time resolution of the temperature measurements in the blowdown pipe. Nevertheless, the results demonstrated that simulations with variable effective momentum which is selected within the experimental uncertainty have provided reasonable agreement with test data on transient temperature distribution in the pool. In order to reduce uncertainty in both experimental data and EHS/EMS modeling, additional tests and modifications to the experimental procedures and measurements system in the PPOOLEX facility were proposed. Pre-test simulations were performed to aid in determining experimental conditions and procedures. Then, a new series of PPOOLEX experimental tests were carried out. A validation of EHS/EMS models against MIX-01 test is presented in this report. The results show that the clearing phase predicted with 3D drywell can match the experiment very well. The thermal stratification and mixing in MIX-01 is also well predicted in the simulation.

  • 137.
    Huang, Zheng
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Numerical Investigations on Debris Bed Coolability and Mitigation Measures in Nordic Boiling Water Reactors2019Doktoravhandling, med artikler (Annet vitenskapelig)
    Abstract [en]

    This thesis is aiming at coolability assessment of particulate debris beds formed in hypothetical severe accidents of Nordic boiling water reactors (BWRs) which may employ either lower drywell flooding or control rod guide tubes (CRGT) cooling as severe accident management strategies. For this purpose, quench and cooling limit (dryout) of debris beds after their formation from fuel coolant interactions were investigated by numerical simulations using the MEWA code. The thesis works consist of four following tasks: (i) validations of the computational tool (the MEWA code) against latest experiments: POMECO-FL and Tutu tests for friction laws, POMECO-HT tests for dryout, and PEARL tests for quench; (ii) Assessment of the long-term coolability of prototypical debris beds with varied characteristics (e.g. height, shape); (iii) studies on quench of prototypical debris beds; (iv) effectiveness analysis of mitigation measures dedicated to Nordic BWRs in term of debris coolability enhancement.

    The comprehensive validations indicate that the MEWA code is a credible and computationally-efficient tool to simulate the two-phase thermal hydraulics of particulate debris beds under both thermal equilibrium and non-thermal equilibrium conditions. Comparisons of the predicted results with experimental data showed a satisfactory agreement, and key phenomena were reproduced.

    Simulation for the prototypical debris beds of the cylindrical, conical and truncated conical shapes showed that the beds’ heights were significantly affecting their coolability, and the values of their dryout power density were roughly inversely proportional to their heights regardless of shapes. Such a relationship was correlated based on the simulation results, which can be employed to guide design and operation of relevant experiments. The impacts of bed’s shape on coolability can be characterized by three factors: multidimensionality and contour surface area of debris bed, as well as uniformity of bed’s shape. An increase in uniformity can improve coolability, since it promotes uniform distributions of temperature and void fraction.

    Simulations for an ex-vessel heap-like debris bed for a reference Nordic BWR showed that the quench front propagated in a multi-dimensional manner. It was found that the upper region adjacent to the centerline of the bed was subject to a higher risk of remelting. It is also found that the oxidation of the residual Zr in the corium had a great impact on coolability of the debris bed due to the release of reaction heat and H2. Therefore, it is crucial to lower the temperature of the whole bed to avoid escalation of oxidation.

    Based on insights from previous studies, several coolability enhancement concepts were proposed as mitigation measures, including downcomer, bottom injection and CRGT cooling. Simulations demonstrated that all the three measures were effective to improve the debris bed coolability. An embedded downcomer increases the cooling capacity by inducing an extra downward flow of water and providing a preferential exit path for steam. Compared with top-flooding, water injection from bottom was predicted to be more efficient to quench a debris bed, since the water inflow was not hindered by the upward flow of steam and therefore could infiltrate the whole bed quickly. The CRGT cooling strategy, applying coolant injection to a particulate debris bed in the lower head of PRV, was proved to be practically feasible to quench the in-vessel debris bed. However, a special attention should be paid to the side effect of Zr oxidation, since it may deteriorate the quenching process and lead to an uncoolable state as a result of release of considerable heat and H2. Therefore, it is essential that the in-vessel debris bed is sufficiently cooled to such an extent during its formation that substantial oxidation would not occur when the CRGT cooling is applied.

  • 138.
    Huang, Zheng
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Numerical investigation on quench of an ex-vessel debris bed at prototypical scale2018Inngår i: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 122, s. 47-61Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    This paper is concerned with the coolability of the heap-like debris beds formed in the cavity of a Nordic-type boiling water reactor (BWR) during a postulated severe accident. A numerical simulation using the MEWA code was performed to investigate the quenching process of the ex-vessel debris bed at post-dryout condition upon its formation. To qualify the simulation tool, the MEWA code was first employed to calculate the quenching tests recently conducted on the PEARL facility. Comparisons of the simulation results with the experimental measurements show a satisfactory agreement. The simulation for the debris bed of the reactor scale shows that the heap-like debris bed flooded from the top is quenched in a multi-dimensional manner. The upper region adjacent to the centerline of the bed is the most difficult for water to reach under the top-flooding condition, and thus is subject to a higher risk of remelting. The oxidation of the residual Zr in the corium has a great impact on the coolability of the debris bed due to (i) large amount of reaction heat and the subsequent positive temperature feedback, (ii) the local accumulation of the produced H2 which may create a “steam starvation” condition and suppresses the oxidation. As possible mitigation measures of oxidation, the effects of bottom-flooding and bypass on quench were also investigated. It is predicted that the debris bed becomes more quenchable with water injected from the bottom, especially for the case with the floor partially flooded in the center. A bypass channel embedded in the center of the debris bed can also promote the quenching process by providing a preferential path for both steam escape and water inflow.

  • 139.
    Huang, Zheng
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    On the quench of a debris bed in the lower head of a Nordic BWR by coolant injection through control rod guide tubesManuskript (preprint) (Annet vitenskapelig)
  • 140.
    Huang, Zheng
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    On the quench of a debris bed in the lower head of a Nordic BWR by coolant injection through control rod guide tubes2019Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 351, s. 189-202Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Since the reactor pressure vessel (RPV) of a typical BWR features a lower head that is penetrated by a forest of control rod guide tubes (CRGTs), coolability of the debris bed formed in the lower head during a severe accident can be realized by coolant injection through the CRGTs (so-called "CRGT cooling"). This paper is concerned with performance assessment of such CRGT cooling system, whose heat removal capacity is determined by two mechanisms: (i) heat-up and boiling of coolant inside the CRGTs; and (ii) evaporation of coolant which reached the top of the debris bed from CRGTs (top flooding). For this purpose, analyses were accomplished by coupling the COCOMO and RELAP5 codes, which simulate the quenching process of the debris bed and the coolant flow inside the CRGTs, respectively. An analysis was first carried out for a unit cell with a single CRGT, whose decay heat removal was limited by heat conduction from debris to the CRGT wall. The simulation indicated that without top flooding, though the temperature of the unit cell was eventually stabilized by the cooling of the CRGT wall, remelting of metallic debris (Zr) in the peripheral region was unavoidable due to low conductivity of corium. Boiling in the CRGT was not only beneficial to heat transfer, but also contributing to a flat axial temperature profile. Given the nominal flowrate of the CRGT cooling, the coolant was not completely boiled off in the CRGT, and therefore the remaining liquid water at the outlet of the CRGT was available for top flooding of the debris bed. The subsequent simulation including the top flooding showed that the debris bed was rapidly quenched without any remelting. However, the top flooding may have a side effect which was Zr oxidation risk at high temperature, leading to production of reaction heat and H-2. Finally analyses were performed for prototypical cases for a reference Nordic BWR, and the results implied that the CRGT cooling could be used as a promising strategy for severe accident mitigation. It is critical that the debris bed is sufficiently cooled down during its formation so that the oxidation risk is eliminated when the CRGT cooling is applied.

  • 141.
    Huang, Zheng
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Performance evaluation of passive containment cooling system of an advanced PWR using coupled RELAP5/GOTHIC simulation2016Inngår i: NUCLEAR ENGINEERING AND DESIGN, ISSN 0029-5493, Vol. 310, s. 83-92Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Motivated to investigate the thermal hydraulic characteristics and performance of a passive containment cooling system (PCS) for a Generation III pressurized water reactor (PWR), a coupled RELAP5/GOTHIC model was developed, which was then employed to simultaneously simulate the transient responses of the PCS and the containment during a large break loss of coolant accident of the reactor. The results show that the PCS is capable of lowering the containment pressure to an acceptable level for a long period (up to 3 days). In a separate-effect study, it was found that the height of the PCS loop plays an important role in determining the flow characteristics and heat removal performance of the PCS. Within the range of the considered loop heights, phase change occurs in the riser of the loop after the height exceeds a specific value (between 13 m and 15 m), below which only single-phase flow takes place. With increasing height of the loop, the heat removal capability increases monotonically at first; however, it is no longer sensitive to the height after two-phase flow appears. Finally, a feed-and-bleed operation for the cooling tank of the PCS was proposed as an enhancement measure of the heat removal capacity, and the simulation results show it further mitigates the accident. Moreover, a simplified analytical model is developed to predict the impact of the feed-and-bleed flowrate on the PCS performance, which can be used in engineering design.

  • 142.
    Huang, Zheng
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Performance of a passive cooling system for spent fuel pool using two-phase thermosiphon evaluated by RELAP5/MELCOR coupling analysis2019Inngår i: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 128, s. 330-340Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    In the aftermath of the Fukushima Daiichi nuclear accident, a great concern has been raised about enhancing the inherent safety of a spent fuel pool (SFP). A passive cooling system using two-phase thermosiphon loops was concerned in this paper. A RELAP5/MELCOR coupling interface was developed, aiming at simultaneously simulating the transient behaviors of the SFP (by MELCOR) and the passive cooling system (by RELAP5). First the RELAP5 model of the thermosiphon loop was qualified against an experiment of a prototypical scale. Comparisons between the experiment and predictions show a good agreement. MELCOR standalone calculations for both station blackout (SBO) and loss of coolant accident (LOCA) without the passive cooling system demonstrate severe degradation of fuel rods. In contrast, for the SBO accident, the coupling simulation shows that the passive cooling system can effectively remove the decay heat, thus keeping fuel rods intact. As for the LOCA scenario, it is more challenging for the passive cooling system due to: (i) the heat transfer power is low during the drainage of water since the natural circulation of steam is blocked by the residual water at the bottom, leading to unavoidable heat-up and oxidation of fuel cladding; (ii) the heat transfer coefficient between steam and the evaporator is very small, which consequently may require a larger heat transfer surface area. Nevertheless, the heat transfer power substantially increases after the pool is emptied and natural circulation is established. The decay heat can be removed by steam convection, thus maintaining the mechanical integrity of fuel rods and stabilizing the fuel temperature eventually. It is also observed that H 2 production is undesirably promoted because the steam supply is enhanced. However such adverse effect can be diminished by increasing the thermosiphon loops number.

  • 143.
    Huang, Zheng
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Validation and application of the MEWA code to analysis of debris bed coolability2018Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 327, s. 22-37Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    This paper was first aimed at validating the MEWA code against experiments for two-phase flow and dryout in particulate beds, and then investigating the coolability of ex-vessel debris beds with cylindrical, conical and truncated conical shapes assumed to form under severe accident scenarios of a boiling water reactor. The validation was mainly performed against the POMECO-FL and POMECO-HT experiments carried out at KTH for investigating frictional laws and coolability limit (dryout) of particulate beds, respectively. The comparison of the experimental and numerical results shows that the MEWA code is capable of predicting both the pressure drop of two-phase flow through porous media and the dryout condition of various stratified beds. While the coolability of a one-dimensional homogeneous debris bed is bounded by counter-current flow limit (CCFL), the coolability of a heap-like debris bed can be improved due to lateral ingression of coolant in a multi-dimensional geometry. The simulations showed that the dryout power density of a prototypical debris bed was roughly inversely proportional to the bed's height regardless of the bed's shape. The impacts of a debris bed's features on coolability are manifested in three aspects: multidimensionality and contour surface area of the bed, as well as the uniformity of its shape. The contour surface area is defined as the interface between debris bed and water pool, and its effective value depends on the surface orientation that determines the amount of water ingress and vapor escape. The perfect uniformity in bed's shape as cylindrical bed results in even distributions of temperature and void fraction. The dryout power density was also predicted to be strongly correlated to the uniformity of bed's shape. The MEWA simulation also predicted that coolability was improved by an downcomer embedded in the center of debris bed. The efficiency of such enhancement was largely determined by the downcomer's length, whose optimal value was obtained in simulation.

  • 144.
    Huang, Zheng
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Thakre, Sachin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Validation of the MEWA code agsinst POMECO-HT experiments and cool ability analysis of stratified debris BEDS2015Inngår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, 2015, Vol. 4, s. 3279-3291Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Motivated by qualification of the MEWA code for coolability analysis of debris beds formed during severe accidents of light water reactors, the present work presents a validation of the code against the experimental data obtained on the POMECO-HT facility for investigation of two-phase flow and heat transfer limits in particulate beds with various characteristics. The volumetrically heated particulate beds used in the POMECO-HT experiment are packed in various configurations, including homogeneous bed, radially stratification, triangular stratification, axial stratification, and multi-stratification. To investigate coolability enhancement by bottom-fed induced natural circulation, a downcomer is employed. Besides, the influence of the interfacial drag is also studied. The results show that simulation results of the MEWA code is overall comparable with the experimental data in term of dryout conditions of the particulate beds. For the 1-D top-flood case, the dryout heat flux is mainly determined by counter-current flow limit. While for certain cases the multidimensionality may help to break CCFL. Besides, the debris bed’s coolabiltiy can be significantly improved due to the natural circulation flow from the bottom induced by using downcomer. The interfacial drag affects the coolability by means of varying the pressure field inside the bed. For the top-flood case, the dryout condition deteriorates since the vapor and coolant flow reversely and thus the interfacial drag increases the flow resistance. Whereas for the bottom-fed case, the dryout heat flux rises remarkably when considering the interfacial drag, because the vapor and coolant flow in the same direction and the interfacial drag helps to pull coolant upward from the bottom.

  • 145.
    Hultgren, Ante
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Gallego-Marcos, Ignacio
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Simulation of Large Scale Erosion of a Stratified Helium Layer by a Vertical Air Jet using the GOTHIC Code2014Konferansepaper (Fagfellevurdert)
    Abstract [en]

    In case of a severe core degradation in a Light Water Reactor (LWR), significant amount of hydrogen can be produced posing a risk of hydrogen burning and detonation. Reliable prediction of hydrogen build-up, stratification, and mixing in the containment is of paramount importance since the phenomena affect hydrogen distribution in the containment. In this paper, we present a modeling approach using the GOTHIC code. The simulation results were compared against experimental data from the ST1-7 experiment performed in the PANDA facility at the Paul Scherrer Institute (PSI). The ST1-7 experiment consists of an air jet impingement onto a stratified helium layer. The modelling approach uses coupled volumes to introduce in each region of the computational domain (i) adequate mesh resolutions to resolve the gradients of the flow and (ii) appropriate turbulence models in order to resolve locally dominant flow structures. With the adaptive mesh, only about 7400 cells for the 2 PANDA vessels (4 m diameter by 8 m in height cylinders with an interconnecting pipe) is enough to provide reasonably accurate results. We found that using the k-epsilon standard model for the jet region and the mixing length model for the rest of the domain, has provided remarkably good agreement with the experimental data. The erosion of the helium stratified layer before and after the air injection is discussed in detail.

  • 146.
    JARA HEREDIA, DANIEL
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    DNB and Void FractionTRACE Model Validation  with PSBT Experiments2011Independent thesis Advanced level (degree of Master (Two Years)), 20 poäng / 30 hpOppgave
  • 147.
    Jaromin, Maria
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Kozlowski, Tomasz
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Anglart, Henryk
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorteknologi.
    Sensitivity Analysis of CFD Two-Phase Flow in a BWR Fuel Bundle2010Konferansepaper (Fagfellevurdert)
    Abstract [en]

    The goal of this work was to perform an investigation of the sensitivities of the average outlet void fraction and flow pattern prediction in the fuel assembly on important model parameters. The model sensitivities were examined on the following groups of parameters: numerical parameters (e.g. mesh discretization), operating conditions (input power, outlet pressure, mass flow rate and inlet temperature) and closure models for the interphase momentum (e.g. drag, virtual mass, lift, turbulent dispersion, wall lubrication). The average outlet void fraction was evaluated by the CFD code for the steady-state and model sensitivities were identified.

    A significant sensitivity was recognized especially for changes of inlet temperature of coolant and system pressure. Moreover, the turbulent dispersion force and the wall lubrication force had a significant influence on vapor and liquid velocity profiles and also on the distribution of void at the outlet of the fuel assembly.

  • 148.
    Jasiulevicius, Audrius
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Sehgal, Balraj
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Experimental investigation of debris bed quenching with non-condensable gas release2006Inngår i: Energetika, ISSN 0375-8842, Vol. 3, s. 1-8Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    A series of experiments were performed at the POMECO (Porous Medium Coolability) facility at the Nuclear Power Safety Division of the Royal Institute of Technology in Stockholm, Sweden. During the experiments, quenching of uniformly heated debris beds of different porosity with air injection from the bottom was investigated. These sets of experiments extend the earlier experiments performed on the POMECO facility and are applicable to ex-vessel coolability phase of a postulated nuclear power plant severe accident in which the attack on concrete generates non-condensable gases which traverse the debris bed. The beds had been heated up to ∼500 °C before a water pool was established on top of the bed. Air was added from the bottom at different flow rates. The results obtained for quench rates and flooding limits showed that for a very low porosity beds (ε = 0.26), the flooding limit is reached with modest air flow rates, while for a usual porosity (ε = 0.38) bed very high air flow rates would be required. The analysis of the experiments showed that the flooding correlations of Marshall-Dhir and Wallis could describe the measured data.

  • 149. Jeltsov, Marti
    et al.
    Cadinu, F.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kööp, Kaspar
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, P.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    An Approach to Validation of Coupled CFD and System Thermal-Hydraulics Codes.2011Konferansepaper (Fagfellevurdert)
  • 150.
    Jeltsov, Marti
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Grishchenko, Dmitry
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Validation of Star-CCM plus for liquid metal thermal-hydraulics using TALL-3D experiment2019Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 341, s. 306-325Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Computational Fluid Dynamics (CFD) provides means for high-fidelity 3D thermal-hydraulics analysis of Generation IV pool-type nuclear reactors. However, to be used in the decision making process a proof of code adequacy for intended application is required. This paper describes the Verification, Validation and Uncertainty Quantification (VVUQ) of a commercial CFD code Star-CCM + for forced, natural and mixed convection regimes in lead-bismuth eutectic (LBE) coolant pool flows. Code qualification is carried out according to an iterative VVUQ process aiming to reduce user effects. Validation data is produced in TALL-3D experimental facility - a 7 m high LBE loop featuring a 3D pool-type test section. Accurate prediction of mutual interaction between thermal stratification and mixing in the pool and the loop dynamics requires 3D analysis, especially during natural circulation. Solution verification is used to reduce the numerical uncertainty during code validation activities. Sensitivity Analysis (SA) is used to identify the effect of the most influential uncertain input parameters (UIPs) on numerical results. Two new visualization methods are proposed to enhance interpretation of the SA results. Dedicated experiments are performed according to the SA results to reduce the uncertainties in the most important UIPs. Automated calibration method for large CFD models is tested and demonstrated in combination with manual calibration using detailed temperature profile measurements in the pool. Calibration reveals the deficiencies in the modeling of heat losses owing to the presence of thermal bridges and other local effects in thermal insulation that are not explicitly modeled. It is demonstrated that Star-CCM + is able to predict thermal stratification and mixing phenomena in the pool type geometries. The results are supported by an Uncertainty Analysis (UA).

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